ML20196K202
| ML20196K202 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/31/1998 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20196K192 | List: |
| References | |
| NUDOCS 9907080201 | |
| Download: ML20196K202 (48) | |
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s Docket No. 50-245 B17801 l
Millstone Nuclear Power Station, Unit No.1 10CFR50.59 Annual Report January 1,1998 through December 31,1998 j
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s U.S. Nucitar R::gulatory Commission B17801\\ Attachment 1\\Page 1 MILLSTONE UNIT No.1 TABLE OF CONTENTS Section Paae No.
INTRODUCTION 2
PLANT DESIGN CHANGES (PDCR / DCR )
3 PROCEDURE CHANGES 6
TECHNICAL REQUIREMENTS MANUAL CHANGES (TRMCR) 13 FINAL SAFETY ANALYSIS REPORT CHANGES (FSARCR) 16 TECHNICAL SPECIFICATION CHANGES (PTSCR)
None to Report BYPASS / JUMPER TEMPORARY MODIFICATIONS 31 l
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U.S. Nuclear R gulatory Commission B17801\\ Attachment 1\\Page 2 INTRODUCTION None of the plant design changes, procedure changes, technical requirements, jumper-lifted leads-bypasses, tests, or experiments described herein constitute (or constituted) an unreviewed safety question per the criteria of 10CFR50.59.
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U.S. Nucber i bgulatory Commission B17801\\ Attachment 1\\Page 3 PLANT DESIGN CHANGES Plant Desian Chance Number Title DCR M1-97035 Emergency Diesel Generator Air Start Pressure Switch Modification DCN-DM1-00-0156-98 Relocation of Propane Bottle and Piping for House Heating Boilers l
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U.S. Nucl=r R:gul: tory Commission B17801\\ Attachment 1\\Page 4 Plant Desian Chance Number Title DCR M1-97035 Emergency Diesel Generator Air Start Pressure Switch Modification Description of Chanae This modification has been released to operations based on implementation of new pressure switch setpoints and revision of low pressure alarm setpoints.
Valve 1-DGSA-31 was made permanent by this modification. Installation of new pressure indicators PI-33-7 and PI-33-10 and associated tubing between the indicators and the air receiver tanks was partially released to operations in the 1997 reporting period.
Reason for Chanae This modification reduced the possibility of challenging the receiver tank relief valve setpoint by replacing switches PS-33-4 & 7 with more accurate pressure switches and moving the sensing point from the Starting Air header to the receiver tanks. Because the pressure switches will be sensing the receiver tank pressure, each tank must be equipped with two pressure switches to operate the respective compressors.
Safety Evaluation The changes to the Emergency Diesel Generator Starting Air System pressure switches and pressure switch setpoints were evaluated as safe and do not constitute an unreviewed safety question.
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U.S. Nucinar R:gulatory Commission B17801\\ Attachment 1\\Page 5 Plant Desian Chanae Number Title DCN-DM1-00-0156-98 Relocation of Propane Bottle and Piping for House Heating Boilers Description of Chanae This change relocated a new 25 gallon House Heating Boiler pilot ignition propane bottle and regulator from the southeast corner of the Unit 1 Maintenance Shop to the southwest corner of the House Heating Boiler Exhaust Stack.
Reason for Chanae This change was made in order to establish greater than 20 feet distance between the propane bottle and the above ground 15,000 gallon Emergency Diesel Generator fuel oil storage tank in accordance with NFPA 30. The previously installed 100 gallon bottle was removed prior to installation of Temporary Modification,1-98-006, for the EDG fuel oil storage tank and the new 25 gallon propane bottle was installed to support continued operation of the House Heating Boilers.
Safety Evaluation The location of the propane tank does not violate any of the Fire Protection Program requirements since it is located in accordance with NFPA 30 requirements for adequate separation, and thus does not adversely affect safety related equipment. The newly installed propane bottle is restrained to ensure that the bottle will not be subject to movement during flooding or high winds.
Additionally, barriers are installed to prevent the bottle from being subject to contact by vehicle traffic.
The change to the location of the propane bottle was evaluated as safe and does not constitute an unreviewed safety question.
i U.S. Nucl:ar R:gul: tory Commission B17801\\ Attachment 1\\Page 6 PROCEDURE CHANGES Procedure Number Title SPROC 97-1-21 Gas Turbine Generator Air Start Test and Generator Rotor Balance Procedure SPROC 97-1-41 Draining Fire Water Tank M7-6A MP 790.4 Control of Heavy Loads OP 328C Refuel Platform Main Hoist Operation HPP-50385-1 Blackness Testing i
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U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 7 Procedure Number Title SPROC 97-1-21 Gas Turbine Generator Air Start Test and Generator Rotor Balance Procedure Description of Chanae SPROC 97-1-21," Gas Turbine Air Start Test and Generator Rotor Balance Procedure," has two objectives. The first is to demonstrate taat the air receiver contains sufficient inventory for three cold starts without recharging and the second is to provide instructions to balance the electric generator rotor. These two activities have been combined because they both require a minimum of three starts. By combining them, wear and tear caused by starting is minimized.
Reason for Chanae As stated in UFSAR section 8.3.1.1.5.2, air for starting is supplied from a 250 psig air receiver which provides sufficient capacity for three successive start attempts without recharging. Review of the pre-operational test number D-9 indicated that the initial start was performed with the air receiver at 250 psig and the subsequent starts were performed at 213 psig and 178 psig. A review of the setpoints that control the start and stop of the starting air compressor indicates that the air receiver pressure is maintained between 200 and 235 psig.
Therefore, since the Ges Turbine Generator (GTG) starting air receiver pressure has consistently been maintained below 250 psig, the pre-operational test was invalid.
Data taken from the GTG surveillance tests, shows that the electric generator vibration amplitude is 0.4 to 0.5 ins /sec at a load of 7.5 MW and 0.6 to 0.7 ins /sec at a load of 10.7MW, the frequency of this vibration is 3600 CPM.
Analysis shows that this vibration can be reduced by balancing and since these levels are considered to be excessive, balancing will be undertaken. Balancing will be achieved by the addition of a trial weight to a vee-grove at one or both ends of the rotor. The GTG is then started and the vibration amplitude and phase angle measured, this information is used to calculate the correct location and size of the final permanent balance weight. Subsequent starts and weight adjustments may be required to fine tune the balance operation.
Safety Evaluation SPROC 97-1-21 " Gas Turbine Air Start Test and Generator Rotor Balance Procedure" is performed while the G'i G is " Inoperable" and " Unavailable" for the mitigation of a COLD SHUTDOWN and Loss of Normal Power accident.
Additionally, the balancing operation will require the GTG to be paralleled with off-site power and loaded, via safety bus 14G, to confirm that the rotor behaves
U.S. Nucl:ar R:gulatory Commission B17801\\ Attachment 1\\Page 8 as predicted when at normal operating temperature. Under these circumstances the GTG cannot be considered to be separated from other safety systems, however, the activity does not change the configuration of the plant nor require changes to operating procedures. The EDG will remain "Available" throughout the test and has the capacity to carry the COLD SHUTDOWN, LNP accident loads. Therefore, this procedure has been reviewed and determined to be safe and does not involve an unreviewed safety question.
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B17801\\ Attachment 1\\Page 9 Procedure Number Title SPROC 97-1-41 Draining Fire Water Tank l
M7-6A Description of Chanae included in this SPROC is guidance to physically drain the fire water tank, align the Unit One Fire Pumps so that they do not recirc to the 'A' tank, and assure minimum flow requirements for the Unit 2 Electric Fire Pump were maintained had the pump been required.
Reason for Chance SPROC 97-1-41 provided direction on how to drain the 'A' Fire Water Storage Tank to allow for intemal inspection of the tank in support of activities required for Bypass / Jumper 1-97-050 (pg. 32).
Safety Evaluation The evaluation had determined that the evolutions required to drain the 'A' tank along with the modification of the Unit 2 Fire Pump minimum flow recirc line and two test procedures were safe and did not create an unreviewed safety question.
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U.S. Nucl2ar R:gulatory Commission B17801\\ Attachment 1\\Page 10 Proegdure Number Title MP 790.4 Control of Heavy Loads Description of Chanae Mechanical / Structural considerations relative to the potential load drop consequences of the items being handled within the safe load paths described in the attachments of Maintenance Procedure MP 790.4, Revision 9," Control of Heavy Loads" were modified by this change.
Reason for Chanae Pursuant to the commitment stated in License Event Report (LER) 96-016-00,
" Heavy Loads Suspended over Irradiated Fuel in the Spent Fuel Pool", the safe load paths specified in maintenance procedure MP 790.4, Revision 8 were reviewed and found to need revision.
Safety Evaluation The movement of plant components specified in ina attachments to MP 790.4 is performed using normal rigging / handling practicos and equipment, and required plant support systems that are to be operated according to existing plant procedures. All of the design parameters required to assure that stored spent fuel is maintained in a safe, coolable, subcritical configuration remain within their j
acceptance limits. Therefore, the operation is safe and does not constitute an unreviewed safety question.
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U.S. Nucl ar R:gulatory Commission B17801\\ Attachment 1\\Page 11 Procedure Number Title OP 328C Refuel Platform Main Holst Operation Description of Chance This change allowed refuel mast and blade guide movement within the spent fuel pool without requiring secondary containment or standby gas treatment systems to be operable.
Reason for Chanae This change was required in support of the Millstone Unit 1 Boraflex Surveillance Program Blackness Testing operations within the spent fuel pool. This one time procedure change permitted refuel mast and blade guide movement in the spent fuel pool without secondary containment or standby gas treatment systems being operable.
Safety Evaluation The procedure change was safe and was not an unreviewed safety question.
The movement of the telescopic fuel handling mast and blade guide within the spent fuel pool has been evaluated and the results demonstrate that the radiological consequences of a potential load drop are bounded by Chapter 15.8,
" Fuel Handling Accident."
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U.S. Nucl:ar Rcgulatory Commission B17801\\ Attachment 1\\Page 12 Procedure Number Title
' HPP-50385-1
. Blackness Testing Description of Chanae Blackness testing'of selected Millstone Unit 1 Spent Fuel Pool storage locations was performed by HOLTEC Intemational. The test lowers a neutron source and detector into selected empty storage locations. Based on the response of the neutron detector to the source, the condition of the Boraflex material in the fuel storage racks can be determined. No fuel movement was performed for blackness testing. The testing was performed following guidance provided by procedure HPP-50385-1 after minor editorial changes we;e made.
Reason for Chance Procedure HPP-50385-1 is a HOLTEC procedure that required approval prior to use at Millstone. The procedure supported the Millstone Unit 1 Boraflex Surveillance Program.
Safety Evaluation The performance of blackness testing on selected locations of the Millstone Unit 1 fuel storage racks did not alter existing malfunctions or accidents, nor did it create new malfunctions or accidents. The blackness test equipment used can result in a maximum load drop of 160 pounds from 3 feet above the storage racks. Such a drop would not preven; the racks from performing their safety function and no fuel pin failures would occur for irradiated fuel stored in the racks. The neutron source used would not change the K-eff of the pool, hence there was no impact to the margin of safety. The blackness test procedure was evaluated to be safe and was not an unreviewed safety question.
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U.S. Nucl=r Rngulatory Commission B17801\\ Attachment 1\\Page 13 TECHNICAL REQUIREMENT MANUAL CHANGES Technical Reauirement Manual Chanae Reauest Number Title TRM 273-7.2 Fire Protection Technical Requirements
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TRM 273-7.4 Control Room Habitability Technical Requircaments i
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U.S. Nucl=r R:gulatory Commission B17801\\ Attachment 1\\Page 14 Technical Reauirement Manual Chanae Reauest Number Title TRM 273-7.2 Fire Protection Technical Requirements Description of Chanae This change consisted of removing one of two existing smoke detectors from the Fire Pump House and replacing it with a heat detector.
Reason for Chanae l
The purpose of the change is to preclude the discharge of the associated Halon Fire Protection system prematurely on smoke only. The new detector is rated for 225 F, with a rate compensating feature. The installation of this new heat detector, in conjunction with the existing smoke detector, would require both smoke and heat to be present before the Halon system would discharge into the enclosure.
The use of a heat detector is preferred because the possibility exists for the Diesel Fire Pump Engine to, exhaust minor amounts of smoke into the room exposing both existing smoke detectors. The heat detector is not sensitive to smoke thus assuring that the Halon system is not subject to trigger on a single change in environment.
Safety Evaluation The change will not increase the probability of an accident or malfunction, affect accident mitigation, or increase the consequences of an accident. The change is safe and doos not create an unreviewed safety question.
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U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 15 Technical Reauirement Manual Chance Reauest Number Title TRM 273-7.4 Control Room Habitability Technical Requirements Description of Chanae This change affected the Unit 1 Technical Requirements Manual Section 273-7.4 for Control Room Habitability. The specific changes to the TRM will; (1) Define operability of Control Room Habitability for Unit One, (2) Delineate how Control Room Habitability will be demonstrated including frequency of required surveillances and actions required in the event that operability criteria cannot be met, and (3) Add the action statement to existing Self-Contained Breathing Apparatus (SCBA) requirements.
Reason for Chanae Design Changes implemented under DCR M1-96056 and DCR M1-97019 were the bases for these TRM changes, j
During periods in which control room habitability is declared inoperable, the TRM requirement to place the unit in a safe, shutdown condition ensures that a radiological release from MP2 or MP3 will not prevent the MP1 operators from maintaining their unit in a safe, stable state. The change to add the action statement is to ensure that the proper number of SCBAs and/or air cylinders is i
maintained.
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4 Safety Evaluation The change was evaluated to be safe and does not create an unreviewed safety question. The control room habitability requirements imposed by the TRM char ;;e ensure that MP1 Operations will be able to effectively manage MP1, regardless of the status of MP2 and MP3.
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U.S. Nuclic.r Regulatory Commission B17801\\ Attachment 1\\Page 16 FSAR Chanae Reauest FS.AR Chanae Reauest Number Title 96-MP1-5 Materials Selection and Fabrication 97-MP1-4 DCR M1-96074 Removal of FST/CUFST 97-MP1-18 DCR M1-97036 -
Emergency Diesel Generator Fuel Supply Header Pressure Increase 97-MP1-20 DCR M1-96056 - High Radiation Trip of Control Room Ventilation 97-MP1-29 PDCR 1-02 Spent Fuel Pool Bulk Temperature Monitoring System 97-MP1-33 Amendment 102 -
Response Time Testing 98-MP1-2 Change to the operation of the Fuel Pool Filter and Demineralizer 98-MP1-3 Impact Load of MP1 Refueling Platform Fuel Grapple Mast and Fuel Assembly of Spent Fuel Pool 98-MP1-6 Fuel Handling Accident 98-MP1-7 Radiological Effluent Monitoring and Off-Site Dose Calculation Manual 98-MP1-9 Steam Tunnel Damper FSAR Correction
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.s U.S. Nucirr R:gulatory Commission B17801\\ Attachment 1\\Page 17
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98-MP1-12 Revision to Millstone Unit 1 UFSAR Section 11.3.2.4, " System Operation" 98-MP1-15 Revision to Chapter 15 Accident Analysis i
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U.S. Nuclear Rngulatory Commission B17801\\ Attachment 1\\Page 18 FSAR Chanae Reauest Number Title 96-MP1-5 Materials Selection and Fabrication Description of Chance A footnote was added to Table 3.2-4 and section 6.1.1.1, that read, "The original plant piping specification, MPC-M1-P1B, specifies ' Pipe Code AL-1', which is 150 lb. rated aluminum piping. Pipe flanges meeting this specification, fabricated from B247 alloy 3003 material, are qualified via design by analysis using ASME BPVC Section Vill, Division 2,1968 edition criteria."
Reason for Chanae Certain 150 lb. Aluminum flanges at Millstone Unit 1 lacked the proper documentation to demonstrate their adequacy for service in the systems in which they were installed.
Safety Evaluation The use of 150 lb. Aluminum flanges in the safety-related piping systems at Millstone Unit 1 is safe. This design does not add or change any lines or components in the Fuel Pool Cooling, Condensate, Condensate Storage and j
Transfer, and Rad Waste High Purity systems. These flanges will not contribute i
to any previously analyzed accident or it's consequences, nor will they contribute to any new accident in addition to those already analyzed. The evaluation concluded that the FSAR change is not an unreviewed safety question.
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B17801\\ Attachment 1\\Page 19 -
FSAR Chance Reauest Number Title 97-MP1-4 DCR M1-96074 Removal of FST/CUFST Description of Chanae UFSAR Sections 11.2,11.4 and table 11.2-2 were changed to incorporate the modifications to the system description resulting from DCR M1-96074. This I
modification provided instructions and documentation to isolate the Clean-up Filter Sludge Storage Tank (CUFST) and the Filter Sludge Storage Tank (FST) via the installation of pipe caps on various input lines to the tanks. This modification also physically removes the CUFST and FST and their associated piping, instrumentation, components and supports.
DCR M1-96074 has been partially released to operations based on removal of :
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- 1. Electrical circuits and associated raceways for Decant Pump M6-52
- 2. Electrical components FSO-6-216, FSO/FCV-6-113 & 114
- 3. Electrical circuits and associated raceways for CUFST and FST components LT-6-23 & 24, PS-6-30, FSO/FCV-6-90 & -501, FSO/FCV-1901-157A&B, and PS-10-6A&B.
This partial release was documented in the Millstone Unit 1 1997 Annual Operating Report.
Reason for Chanae The Clean-up Filter Sludge Storage Tank (CUFST), M6-1, and the Filter Sludge Storage Tank (FST), M6-2, had been operationally removed from service due to leakage, inspection of these tanks had revealed that the tanks were beyond repair and should be physically removed and replaced. This change documents the removal of these tanks from the FSAR. The replacement of the CUFST and
~ FST with a new Decant Sludge Tank will be addressed via DCR M1-96060.
j Safety Evaluation The changes made by DCR M1-96074, to remove the Clean-up Filter Sludge
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Storage Tank (CUFST) and the Filter Sludge Storage Tank (FST) will not affect -
any design basis accident or its consequences. It will not contribute to any new accidents beyond those already analyzed. The change was evaluated and I
. determined to be safe and does not present an unreviewed safety question.
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U.S. Nucl:ar R:gulatory Commission B17801\\ Attachment 1\\Page 20 FSAR
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Chanae Reauest Number Title 97-MP1-18 DCR M1-97036 -
Emergency Diesel Generator Fuel Supply Header Pressure Increase Description of Chance UFSAR Section 8.3.1.1.5.1 was changed to incorporate the modification to the system description resulting from DCR M1-97036. This design change implemented the Original Equipment Manufacturer's (OEM) current recommendation to change the Emergency Diesel Generator (EDG) fuel supply end header pressure from approximately 15 psi that is stated per UFSAR Section 8.3, to a 20-30 psi range (35 psi max) to compensate for the maximum allowable differential pressure across the filter which will also be changed, from 10 psid to 7 psid, Sufficient pressurization of the fuel supply header is required to ensure the complete filling of the injection plungers for the EDG fuel system to maintain reliable operation.
Reason for Chanae This change was implemented to bring the Millstone Unit 1 UFSAR and operating practice into compliance with the manufacturer's current recommendations for improved operation of the EDG.
Safety Evaluation The design change required an increase in the initial header end pressure which ensures that sufficient pressure is maintained in the fuel supply header for the complete filling of the fuelinjection pumps and therefore prevent possible engine fuel starvation. The design change has no adverse affect on the probability of occurrence or consequences of previously evaluated malfunction of equipment important to safety and does not create the possibility of a malfunction of a different type than previously evaluated. The design change does not have an adverse affect on the probabi!ity of occurrence or consequences of previously evaluated accidents and does not create the possibility of an accident of a different type than previously evaluated. The design change is concluded to be safe and does not result in an unreviewed safety question.
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U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 21 FSAR Chanae Reauest Number Title 97-MP1-20 DCR M1-96056 - High Radiation Trip of Control Room Ventilation Description of Chance The Habitability Systems, Engineered Safety Feature Systems and HVAC System Sections of the UFSAR were changed to incorporate the modification to the system description resulting from DCR M1-96056 and M1-97019. The design change installed two independent circuits to enable remote trip of Control Room Ventilation fans HVT-11 and HVT-10A and isolation of the Outside Air supply damper 1-HVD-8C. One circuit trips the fans and damper if high radiation conditions are sensed by the radiation detector at the fresh air inlet to the Control Room. The other circuit permits Control Room operators to manually trip the ventilation fans and damper if necessary.
This modification was partially tumed over to operations with all physical work complete. Refer to the Millstone Unit 1 1997 Annual Operating Report.
Reason for Chanae This configuration was selected to ensure that diverse means existed to trip Control Room Ventilation if a Design Basis Accident and subsequent release of radiation were to occur at any of the Millstone Point units.
Safety Evaluation The design change is consistent with Control Room Ventilation trip logic for the other units at Millstone Point, with the exception that Unit One's Control Room ventilation is a single train system that has been previously reviewed and accepted by the NRC. The change is safe and not an unreviewed safety question.
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U.S. Nucinar Regulatory Commission B17801\\ Attachment 1\\Page 22 1
FSAR Chanae Reauest Number Title 97-MP1-29 PDCR 1-02 Spent Fuel Pool Bulk Temperature Monitoring System Description of Chance UFSAR Sections " Spent Fuel Pool Cooling System",9.1.3 and " System Description",9.1.3.3 were changed to incorporate the modification to the system description resulting from PDCR 1-02-96. This modification installed two thermocouples, one on the north wall and one on the west wall approximately centerline of the spent fuel pool. Cables were routed to the Control Room panel (CRP) 904 and connected to existinpi temperature recorder TR-1040-2, channels 10 and 11. The recorder was programmed and the new channels set to alarm at 125 degrees Fahrenheit to actuate the "SHTDN COOLING / FUEL POOL Hi TEMP" annunciator on CRP 904.
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Prior to this modification, temperature measurement and indication was provided on the suction side of the fuel pool cooling pumps and on the discharge side of the fuel pool cooling heat exchangers. Failure of the fuel pool pumps (i.e. loss-of-flow) would have left the plant with no means of monitoring the fuel pool temperature.
- Safety Evaluation
- This modification provides additional instrumentation to the Fuel Pool Cooling System. This modification will not contribute to any previously analyzed accident or its probability or consequences; nor will thir, change contribute to any new accident outside of those already analyzed. The margin of safety, as defined in the basis for any Technical Specifications, is not reduced. There are no unreviewed safety questions associated with this modification.
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U.S. Nucl ar Regulatory Commission B17801\\ Attachment 1\\Page 23 FSAR Chanae Reauest Number Title 97-MP1-33 Amendment 102 -
Response Time Testing Description of Chanae Licensing Amendment 102 made revisions to Technical Specification Section 1.0, " Definitions" and Section 3.1/4.1, " Reactor Protection System" including the associated bases. Amendment 102 modified Millstone Unit No.1 instrument calibration, functional and response time requirements. Also modified were the testing requirements for Resistance Temperature Detectors and thermocouples.
Also, certain definitions were clarified and/or modified using applicable wording j
of NUREG 1433, " Standard Technical Specifications" Revision 1 and Owners approved industry /TSTF Standard Technical Specification Change Traveler (TSTF-64). Additionally, the change relocates the Reactor Protection System logic response time value utilizing the guidance provided by Generic Letter 93-03, " Relocation of Technical Specification Tables of Instrument Response Time Limits," to the Update Final Safety Analysis Report (UFSAR). Amendment 102 clarifies the instrumentation sunteillance requirements, thereby helping to ensure proper testing of safety related components.
Reason for Chanae The changes to the UFSAR are the result of issuance of Amendment 102 as well as correcting an identified discrepancy concerning the drywell vacuum breaker j
valve operation. The UFSAR change reflects the changes to the Millstone Unit 1 l
licensing basis via approval of Amendment 102.
Safety Evaluation These changes to the UFSAR are safe and do not constitute an unreviewed safety question. The UFSAR changes will clarify instrumentation surveillance requirements associated with the RPS and ESF instrumentation to agree with clarifications made within Amendment 102. Also, it corrects a known discrepancy and will not create a plant condition which has not been previously analyzed or create a malfunction or accident not previously considered in the analysis.
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U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 24 FSAR Chanae Reauest Number Title 1
98-MP1-2 Change to the operation of the Fuel Pool Filter and l
Demineralizer 1
l Description of Chanae Condition Report M1-97-2235 identified that the Fuel Pool Filter was not in operation continuously contrary to UFSAR section 9.1.2.2, which states that the water in the pool is continuously filtered and cooled by the fuel pool cooling system. This change clarified the operation of the fuel pool filter and the existing operation of the fuel pool demineralizer.
Reason for Chanae The Fuel Pool Cooling System operating procedure, OP310, does not provide the specifics of the operation of the filter however it does state that the system would be operated as necessary to maintain desired water temperature, quality and clarity. This change to the FSAR updates it to reflect the current use of the Fuel Pool Cooling System.
Safety Evaluation l
The safety evaluation concluded that operation of the fuel pool filter and the demineralizer to maintain pool water clarity and water quality on an as-needed basis as directed by the chemistry sampling procedure, was consistent with the design basis intent of the filter and demineralizer. This change was evaluated as safe and does not create an unreviewed safety question.
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U.S. Nucl:cr R:gulatoiy Commission B17801\\ Attachment 1\\Page 25 FSAR Chanae Reauest Number Title l
98-MP1-3 Impact Load of MP1 Refueling Platform Fuel Grapple Mast and Fuel Assembly of Spent Fuel Pool Description of Chanae it was determined that a single active failure, such as a cable break, could result in a dropped fuel assembly along with portions of the mast falling onto fuel stored in the spent fuel storage racks or in an empty storage rack cell. The additional kinetic energy from the falling mast sections was not included in the design of the spent fuel storage racks. Analyses have been performed to demonstrate that the spent fuel racks remain functional during a potential drop event.
Reason for Chanae This change incorporated the analyses for spent fuel pool storage racks into sections 9.1.2.3 of the UFSAR.
Safety Evaluation All of the design parameters required to assure that stored spent fuel is maintained in a safe, coolable, subcritical configuration remain within their acceptance limits. Radiological consequences due to fuel damage and storage rack damage during movement of new fuel into the spent fuel pool, are bounded by previous analyses. Therefore, the change is safe and does not constitute an unreviewed safety question.
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U.S. Nuclsar R::gulatory Commission B17801\\ Attachment 1\\Page 26 FSAR Chanae Reauest Number Title 98-MP1-6 Fuel Handling Accident Description of Chanae Analyses have been performed to assess the impact of the additional weight of the refueling mast on various fuel handling accident scenarios and the results were incorporated into the FSAR by this change. A failure of the hoist cable would result in not only the fuel assemble falling, but also sections of the refueling mast. This additional weight and resulting kinetic energy was not included in the analysis. Also, the FSAR analysis only addressed a drop within the reactor vessel. Although it could be postulated that this was the limiting condition, analyses had not been previously performed and incorporated into the licensing basis that confirmed this conclusion.
Reason for Chanae The FSAR safety analysis for the Fuel Handling Accident did not account for the weight of the mast in the impact analysis. This change incorporated the analyses for the Fuel Handling Accident into section 15.8 of the UFSAR.
Safetv Evaluation This UFSAR Change updated the discussion of the Fuel Handling Accident. The published acceptance criteria for offsite radiological releases associated with this event per SRP 15.7.4 and NUREG 0612 are that calculated whole-body and thyroid doses at the exclusion area and low population zone boundaries are well within (less than 25%) of the 10CFR100 limits.
The analyses performed demonstrated that the results not only meet the above j
criteria, but that they remain bounded by the existing analysis of record.
This change is safe and la not an unreviewed safety question.
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B17801\\ Attachment 1\\Page 27 FSAR' y
'Chanae Reauest Number Title i
98-MP1-7 Radiological Effluent l
Monitoring and Off-Site Dose Calculation Manual Description of Chanae The Off-Site Dose Calculation Manual (ODCM) was revised to allow the use of dilution water flow from Millstone Units 2 and 3 when discharging liquid radwaste to the_ discharge canal. The FSAR was revised to clarify that dilution water flow is not only from Unit 1.
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Reason for Chanae During extended shutdowns the Unit 1 circulating water pumps may not be l
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Monitoring and Off-Site Dose Calculation Manual (REMODCM) requirements previously would limit the rate of discharge from liquid radwaste tanks into the quarry. Radioactive concentrations in the quany were to be maintained below i
10% of the limits in 10CFR20 using only dilution flow from Unit 1. The legal limit for radioactivity concentration in water released to the environment is applicable at the quarry cut where the water is discharged into Long Island Sound.
Therefore, credit for _ dilution flow from Unit 2 or Unit 3, which are being discharged into the quarry at the time of the Unit 1 discharge, would ensure that the limit on discharges into Long Island Sound is not exceeded. This revision to the FSAR allows Unit 1 to speed up liquid radwaste discharges during extended shutdowns when Unit 1 circulating water pumps are not available.
l Safety Evaluation The change will allow Unit 1 to discharge liquid radwaste using Unit 2 and/or Unit 3 diluting water when dilution flow from Unit 1 is not adequate by itself to maintain radioactivity concentrations in water being discharged to the quarry below the 10CFR20 limits. Unit 1 discharge could exceed the concentrations in 10CFR20 at the point of discharge into the quarry, however, radioactivity concentrations in water being discharged to the Long Island Sound will be below the 10CFR20 limits after mixing with Unit 2 and/or Unit 3 dilution flow. This is consistent with other requirements in the Unit 2 Technical Specifications.
The changes have been reviewed and determined to be safe and does not involve an unreviewed safety question.
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U.S. Nucl ar R:gulatory Commission B17801\\ Attachment 1\\Page 28 FSAR Chanae Reauest Number Title 98-MP1-9 Steam Tunnel Damper FSAR Correction Description of Chanae The design bases for the steam tunnel ventilation system is to isolate with all the same signals as those that initiate the Standby Gas Treatment System. The FSAR description of how this isolation occurred was inaccurate. This change corrected the inaccuracies and clarified the section of the FSAR that described the Steam Tunnel Ventilation System and its operation.
Reason for Chanae The description of the operation of the Steam Tunnel Ventilation System in Section 9.4.8.2 of the FSAR was not accurate. This change corrected those inaccuracies.
Safety Evaluation The implementation of this change is not an unreviewed safety question. This FSAR revision made no changes to the operation of the Steam Tunnel Ventilation System.
4 U.S. Nucl:ar Rtgulatory Commission B17801\\ Attachment 1\\Page 29 FSAR Chance Reauest Number Title 98-MP1-12 Revision to Millstone Unit 1 UFSAR Section 11.3.2.4, " System Operation" Description of Chanae Section 11.3.2.4 of the UFSAR was revised to say that the Off-gas system will.
alarm when release limits are approached. This change allows flexibility in selection of the Steam Jet Air Ejector (SJAE) radiation monitor setpoint.
Reason for Chanae This revision was needed to make the UFSAR consistent with the Radiological Effluent Monitoring and Off-site Dose Calculation Manual (REMODCM).
Although a setpoint of ' ten times the normal discharge rate' was conservative relative to the requirement of the REMODCM, it did not account for the condition when the normal discharge rate may increase to within a factor of ten of the instantaneous noble gas release rate limit.
Safety Evaluation The change was evaluated to ensure that a SJAE radiation setpoint allowed by the UFSAR would not increase the consequences of any accident. The change was also evaluated to ensure that the allowed setpoint would prevent any gaseous radioactive releases from exceeding any technical specification of other regulatory limit.
The change was evaluated as safe and does not present an unreviewed safety question.
l
U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 30 FSAR Chance Reauest Number Title 98-MP1-15 Revision to Chapter 15 Accident Analysis Description of Chance This change removed the reactor related accidents from Chapter 15 of the Millstone Unit 1 UFSAR. The radiological consequences to the Fuel Handling Accident, Section 15.8 of the UFSAR was revised to reflect calculations that use actual source terms and to provide the updated radiological consequences of a bounding Fuel Handling Accident at the Millstone Unit 1 off-site boundary to the Millstone Unit 1 and Millstone Unit 2 control rooms.
Reason for Chanae The change was required as a result of the permanent cessation of operations of Millstone Unit 1 and the permanent removal of fuel from the reactor vessel into the fuel storage pool. The change also updates the radiological consequences of a Fuel Handling Accident at the Millstone Unit 1 off-site boundary and to the Millstone Unit 1 and Millstone Unit 2 control room operators. The revised calculation provides a bounding scenario using actual source terms for calculating the Millstone Unit 1 off-site and to the Millstone Unit 1 and Millstone Unit 2 control room doses from a permanently defueled reactor plant. The actual source terms were based on plant conditions typical for a defueled plant going into decommissioning and the decay history of the fuel in the fuel storage pool.
l Safety Evaluation Removing the reactor-related accidents from chapter 15 of the UFSAR is safe j
because with fuel assemblies permanently removed from the reactor vessel this 1
change causes no risk to public health and safe $.
Revising the radiological consequences of E Fuel Handling Accident in section 15.8 is safe because the revised radiolog%al consequences:
a) are well below the previously analyzed off-site radiological consequences;
' b) do not exceed the SRP 15.7.4 dose limits of 75 Rem to the thyroid and 5 Rem to the Whole Body for exclusion area boundary and low population zone doses; and c) do not exceed the GDC 19 dose limits for the control rooms.
The Change was evaluated as safe and is not an unreviewed safety question.
U.S. Nucl: r R:gul: tory Commission B17801\\ Attachment 1\\Page 31 BYPASS / JUMPER CHANGES Bvoass/ Jumper Number Title 1-97-050 Fire Water Tank Draining Support Equipment 1-98-001 Diesel Exhaust Pipe Temporary Patch 1 98-006 Installation of Temporary EDG Fuel Oil Storage Tank 1-98-007 Temporary House Heating Boiler Fuel Oil Storage Tank 1-98-008 Installation of Temporary GTG Fuel Storage Tank i
m q
U.S. Nuclur Rrgul: tory Commission B17801\\ Attachment 1\\Page 32 Bvoass/ Jumper Number Title 1-97-050 Fire Water Tank Draining Support Equipment Description of Chance This Temporary Modification installed a blind flange, a recirculation hose and an FME cover to support special procedure SPROC 97-1-41 " Draining Fire Water Storage. Tank M7-6A" (pg.9).
Reason for Chanae A blind flange was installed to assure that normal recirculated water could not -
enter the "A" Fire Water Tank while the tank was out of service for draining and inspection. In addition, a valve assembly was installed to accommodate the attachment of a hose with heat tracing to route the discharge of 1-FIRE-89 to the hatch on the "B" Fire Water Tank to allow the Unit 2 Fire Pump to remain OPERABLE during the draining and inspection of the "A" Fire Water Tank. To address the potential for the introduction of FME into the "B" tank hatch, where the recirculating hose was placed, the opening was protected with a FME cover securely attached in place.
Safety Evaluation This safety evaluation determined that the evolutions required to drain the "A" Fire Water Tank and the modification to the Unit 2 Fire Water Pump minimum flow recirc line was safe and does not involve an unreviewed safety question.
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U.S. Nuclear R:gulatory Commission B17801\\ Attachment 1\\Page 33 Bvoass/ Jumper Number Title 1-98-001 Diesel Exhaust Pipe Temporary Patch Description of Chanae The exhaust piping from the Emergency Diesel Generator (EDG) had developed a hole in the elbow before the pipe exits the room through the roof vert!cally into the muffler. The nature of the piping material surrounding the approximately 6-inch diameter hole indicated corrosion / erosion as the cause. A patch was welded onto the existing piping with a circumferential weld around the periphery of the patch to prevent exhaust gases from escaping into the room when the diesel is operating.
Reason for Chanae Temporarily patching the hole along with subsequent inspections for exhaust leakage will ensure the availability of the EDG until the exhaust piping can be replaced permanently.
Safety Evaluation This temporary change does not adversely affect any plant systems or components and is acceptable while Unit 1 is in cold shutdown or refuel modes.
'This Bypass Jumper has been reviewed and determined to be safe and does not involve an unreviewed safety question.
U.S. Nucl:ar Regulatory Commission B17801\\ Attachment 1\\Page 34 Bvoass/ Jumper Number Title
]
1-98-006' installation of Temporary EDG Fuel Oil Storage Tank Description of Chanae This change temporarily modifies the SAR description and operation of the Millstone Unit 1 Emergency Diesel Generator (EDG) as required to support Spent Fuel Pool (SFP) cooling and Millstone Unit 2 Station Blackout (SBO) with (1) all fuel permanently removed from the reactor pressure vessel and (2) no activities which require secondary containment integrity. The specific temporary
)
changes to the design basis of the EDG include:
Changes the credited on-site fuel supply for the EDG by replacing the 23,500 gallon fuel inventory requirement with an above ground storage tank with a 15,000 gallon nominal capacity and means to maintain fuel oil temperature above the cloud point.
Modified the EDG fuel oil transfer system.
Determined the electrical loading on the EDG required to support the equipment associated with the SFP cooling function.
Established the quantity of fuel oil needed to support operation of the EDG for 5 days at the reduced loading.
Specified seismic class 11 requirements for the EDG fuel storage and transfer systema.
Reason for Chanae The existing underground fuel storage tank was required to be emptied prior to exceeding the thirty year life (October 31,1998) as required by the State of Connecticut Department of Environmental Protection (DEP). This change provides a new above ground fuel storage tank for EDG availability to support Millstone Unit 1 SFP Cooling safety function and Millstone Unit 2 SBO program.
Safety Evaluation The only safety related function of the EDG with all fuel permanently removed from the Reactor Pressure Vessel, is to provide SFP Cooling which does not j
require seismic capability. The ability to provide this function is not reduced by a QA Category 1, Seismic Class ll fuel oil supply to the EDG The SBO function is not required to be safety related or seismic. Therefore, the OA Category 1, Seismic Class ll fuel oil supply is acceptable. Therefore, the change was evaluated to be safe and does not constitute an unreviewed safety question.
I i
U.S. Nucl=r R:gul: tory Commission B17801\\ Attachment 1\\Page 35 Bypass / Jumper Number Title 1-98-007 Temporary House Heating Boiler Fuel Oil Storage Tank Description of Chanae This change will provide temporary, above ground fuel oil storage for the House Heating Boile. to support Millstone 1 and 2 heating steam requirements. The change temporarily modifies the SAR description of the site plot plan in the area west of the Gas Turbine Generator. The specific proposed changes to the House Heating Boiler fuel oil storage and transfer system include:
Changes the fuel oil supply for the House Heating Boiler by replacing the two 25,000 gallon underground storage tanks with a single above ground storage tank with a 25,000 gal lon nominal capacity installed in accordance with NFPA requirements.
Modifies the House Heating Boiler fuel oil transfer system - two positive displacement fuel oil transfer pumps provide fuel oil from the new nominal 25,000 gallon tank to the boiler room.
Conducts system testing to verify pipe and component integrity, proper pump operation and ability to deliver suitable flow and pressure to meet operation of all three heating boilers Reason for Chance The existing underground fuel oil storage tanks were required to be emptied prior to exceeding the thirty year life (October 31,1998) as required by the State of Connecticut Department of Environmental Protection. This change provides a new above ground fuel oil storage tank for the Plant Heating Boilers.
Safety Evaluation The House Heating Boiler fuel oil storage and transfer system performs no nuclear. safety related functions. The location of the proposed 25,000 gallon House Heating Boiler above ground storage tank meets the fire protection requirements for separation and clearance between other above ground storage tanks and structures containing safety related equipment. An evaluation of the impact on site flooding of installing the above ground storage tank has been completed and concludes that there is no adverse impact on Millstone Units 1,2, or 3.
Based on the issues addressed above, the proposed changes to the House Heating Boiler fuel oil storage and transfer system is safe and does not constitute an unreviewed safety question.
U.S. Nucl:ar R:gulatory Commission B17801\\ Attachment 1\\Page 36 i
Bypass / Jumper Number Title 1-98-008 installation of Temporary GTG Fuel Storage Tank Description of Chanae This change temporarily modifies the SAR description and operation of the Millstone Unit 1 Emergency Gas Turbine Generator (GTG) such that it is not required to provide credited emergency power to Millstone Unit 1 safety related loads. The specific temporary changes to the design basis of the GTG include:
Changes the quality and seismic requirements of the GTG system from QA category 1, seismic Class I to SBOOA, seismic Class 11.
Changes the credited on-site fuel supply for the GTG by replacing the 35,000 gallon fuel inventory requirement with an above ground storage tank with greater than 4,000 gallon usable inventory.
Design basis external events do not apply.
Reason for Chance The existing underground fuel storage tank was required to be emptied prior to exceeding the thirty year life (October 31,1998) as required by the State of Connecticut Department of Environmental Protection. This change provides a new above ground fuel storage tank for GTG availability to provide emergency
)
power to Millstone Unit 2 as required by the Millstone Unit 2 Station Blackout (SBO) Program.
STATUS - It should be noted that only the concrete tank foundation mat was poured and the 15,000 gallon tank set in place.
Safety Evaluation With all fuel permanently removed from the Reactor Pressure Vessel the EDG is the only on-site emergency power supply required to support the Spent Fuel Pool cooling system, as established by License Amendment 89. Therefore, the only requirement that the GTG must meet is to provide support to the Unit 2 SBO Program. With the replacement of the 35,000 gallon fuel tank by a 4,000 gallon fuel tank the GTG still has the ability to provide sufficient support to the Milhtone 2 SBO loads for eight hours.
The result of this change is to temporarily remove the safety related designation of the GTG system from the licensing basis. This is possible because 10CFR50.2 defines a safety related system as one which will assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; and (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. The GTG is no longer required to fulfil any
i
' U.S. Nuclear R:gulatory Commission B17801\\ Attachment 1\\Page 37
{
of these requirements. There are no accident scenarios in the present
- permanently defueled condition of the plant which require electrical power to
- mitigate the consequences of an accident or to limit the dose to the public.
Therefore there remains no GTG electrical system or component that provides a safety related function, it has been determined that removing the GTG from performing safety related q
functions at Millstone 1 while maintaining Millstone 2 SBO functions is safe and i
does not constitute an unreviewed safety question.
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Docket No. 50-245 B17801 I
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Millstone Nuclear Power Station, Unit No.1 1998 Regulatory Commitment Changes June 1999
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